Preparation and Characterization of 17O-Enriched Alumina for Nuclear Fusion Damage Experiments

Alumina enriched in ¹⁷O was successfully fabricated from aluminum isopropoxide and water containing the ¹⁷O isotope. This material was necessary for an experiment to study the radiation damage expected in alumina exposed to a nuclear fusion reactor environment. The enrichment levels of specimens subjected to different preparation schedules were measured using a nuclear reaction analysis technique. Replacement of the ¹⁷O isotope in the ceramic by atmospheric oxygen occurred readily. Therefore, successful fabrication of suitably enriched alumina specimens required that all processing steps be performed under vacuum or inertgas environments. The optimum fabrication procedure produced enriched alumina specimens of >99.5% of theoretical density, ∽10-μm grain size, and a flexural strength of 280 MPa.     

A Plan for the Development of Fusion Energy

This is the final report of a panel set up by the U.S. Department of Energy (DOE) Fusion Energy Sciences Advisory Committee (FESAC) in response to a charge letter dated September 10, 2002 from Dr. Ray Orbach, Director of the DOE's Office of Science. In that letter, Dr. Orbach asked FESAC to develop a plan with the end goal of the start of operation of a demonstration power plant in approximately 35 years. This report, submitted March 5, 2003, presents such a plan, leading to commercial application of fusion energy by mid-century. The plan is derived from the necessary features of a demonstration fusion power plant and from the time scale defined by President Bush. It identifies critical milestones, key decision points, needed major facilities and required budgets. The report also responds to a request from DOE to FESAC to describe what new or upgraded fusion facilities will best serve our purposes over a time frame of the next twenty years.     

Microstructural and mechanical property changes with aging of Mo-41Re and Mo-47.5Re alloys

The changes in microstructure and mechanical properties of Mo-41Re and Mo-47.5Re alloys were investigated following 1100 h thermal aging at 1098, 1248 and 1398 K. The electrical resistivity, hardness and tensile properties of the alloys were measured both before and after aging, along with the alloy microstructures though investigation by optical and electron microscopy techniques. The Mo-41Re alloy retained a single-phase solid solution microstructure following 1100 h aging at all temperatures, exhibiting no signs of precipitation, despite measurable changes in resistivity and hardness in the 1098 K aged material. Annealing Mo-47.5Re for 1 h at 1773 K resulted in a two-phase αMo + σ structure, with subsequent aging at 1398 K producing a further precipitation of the σ phase along the grain boundaries. This resulted in increases in resistivity, hardness and tensile strength with a corresponding reduction in ductility. Aging Mo-47.5Re at 1098 and 1248 K led to the development of the χ phase along grain boundaries, resulting in decreased resistivity and increased hardness and tensile strength while showing no loss in ductility relative to the as-annealed material.     

Low temperature neutron irradiation effects on microstructure and tensile properties of molybdenum

Polycrystalline molybdenum was irradiated in the hydraulic tube facility at the High Flux Isotope Reactor to doses ranging from 7.2 × 10⁻⁵ to 0.28 dpa at 80 °C. As-irradiated microstructure was characterized by room-temperature electrical resistivity measurements, transmission electron microscopy (TEM) and positron annihilation spectroscopy (PAS). Tensile tests were carried out between −50 and 100 °C over the strain rate range 1 × 10⁻⁵ to 1 × 10⁻² s⁻¹. Fractography was performed by scanning electron microscopy (SEM), and the deformation microstructure was examined by TEM after tensile testing. Irradiation-induced defects became visible by TEM at 0.001 dpa. Both their density and mean size increased with increasing dose. Submicroscopic three-dimensional cavities were detected by PAS even at 0.0001 dpa. The cavity density increased with increasing dose, while their mean size and size distribution was relatively insensitive to neutron dose. It is suggested that the formation of visible dislocation loops was predominantly a nucleation and growth process, while in-cascade vacancy clustering may be significant in Mo. Neutron irradiation reduced the temperature and strain rate dependence of the yield stress, leading to radiation softening in Mo at lower doses. Irradiation had practically no influence on the magnitude and the temperature and strain rate dependence of the plastic instability stress.     

Effect of oxygen on vacancy cluster morphology in metals

The extensive literature on oxygen chemisorption and solubility in metals is briefly reviewed, with special emphasis on the reduction of surface tension associated with oxygen adsorption. A thermodynamic model based on the adsorption equations of Gibbs and Langmuir is developed to determine the relative stability in the presence of oxygen of the void compared to the dislocation loop and stacking fault tetrahedron. Representative calculations are performed for copper, nickel, and austenitic stainless steel. Atomistic and elastic continuum calculations predict that void formation should not occur in most pure face-centered cubic metals during quenching or irradiation. However, the thermodynamic model predicts that oxygen concentrations of 30 to 1000 appm will stabilize void formation in copper, nickel, and stainless steel. Foils of copper and several Fe-Cr-Ni stainless steels containing various amounts of oxygen have been examined with electron microscopy following ion bombardment. The presence of 30 to 1000 appm O resulted in significant amounts of void formation, whereas no voids were observed in low-oxygen specimens, in agreement with the model predictions. Oxygen introduced by ion implantation was more effective in promoting void formation than residual oxygen. Solutes such as phosphorus in stainless steel reduced the effectiveness of oxygen as a void-stabilizing agent. This paper is based on a presentation made in the symposium “Irradiation-Enhanced Materials Science and Engineering” presented as part of the ASM INTERNATIONAL 75th Anniversary celebration at the 1988 World Materials Congress in Chicago, IL, September 25–29, 1988, under the auspices of the Nuclear Materials Committee of TMS-AIME and ASM-MSD     

Tensile and fracture toughness properties of neutron-irradiated CuCrZr

Tensile and fracture toughness properties of a precipitation-hardened CuCrZr alloy were investigated in two heat treatment conditions: solutionized, water quenched and aged (CuCrZr SAA), and hot isostatic pressed, solutionized, slow-cooled and aged (CuCrZr SCA). The second heat treatment simulated the manufacturing cycle for large components, and is directly relevant for the ITER divertor components. Specimens were neutron irradiated at ∼80 °C to two fluences, 2 × 10²⁴ and 2 × 10²⁵ n/m² (E > 0.1 MeV), corresponding to displacement doses of 0.15 and 1.5 displacements per atom (dpa). Tensile and fracture toughness tests were carried out at room temperature. Significant irradiation hardening and plastic instability at yield occurred in both heat treatment conditions with a saturation dose of ∼0.1 dpa. Neutron irradiation slightly reduced fracture toughness in CuCrZr SAA and CuCrZr SCA. The fracture toughness of CuCrZr remained high up to 1.5 dpa (J_Q > 200 kJ/m²) for both heat treatment conditions.     

Aging effects on microstructural and mechanical properties of select refractory metal alloys for space-reactor applications

Refractory alloys based on niobium, tantalum and molybdenum are potential candidate materials for structural applications in proposed space nuclear reactors. Long-term microstructural stability is a requirement of these materials for their use in this type of creep dominated application. Early work on refractory metal alloys has shown aging embrittlement occurring for some niobium and tantalum-base alloys at temperatures near 40% of their melting temperatures in either the base metal or in weldments. Other work has suggested microstructural instabilities during long-term creep testing, leading to decreased creep performance. This paper examines the effect of aging 1100 h at 1098, 1248 and 1398 K on the microstructural and mechanical properties of two niobium (Nb-1Zr and FS-85), tantalum (T-111 and ASTAR-811C) and molybdenum (Mo-41Re and Mo-47.5Re) base alloys. Changes in material properties are examined through mechanical tensile testing coupled with electrical resistivity changes and microstructural examination through optical and electron microscopy analysis.     

Thermal conductivity degradation of ceramic materials due to low temperature, low dose neutron irradiation

The thermal conductivity degradation due to low-temperature neutron irradiation is studied and quantified in terms of thermal resistance terms. Neutron irradiation is assumed to have no effect on umklapp scattering. A theoretical model is presented to quantify the relative phonon-scattering effectiveness of the three dominant defect types produced by neutron irradiation: point defects, dislocation loops and voids. Several commercial ceramics have been irradiated with fission reactor fast neutrons at low temperatures to produce defects. Materials include silicon carbide, sapphire, polycrystalline alumina, aluminum nitride, silicon nitride, beryllium oxide, and a carbon fiber composite. The neutron dose corresponded to 0.001 and 0.01 displacements per atom (dpa) for a 60 °C irradiation and 0.01 and 0.1 dpa for a 300 °C irradiation. Substantial thermal conductivity degradation occurred in all of the materials except BeO following irradiation at 60 °C to a dose of only 0.001 dpa. The data are discussed in terms of the effective increase in thermal resistance caused by the different irradiation conditions. Evidence for significant point defect mobility during irradiation at 60 and 300 °C was obtained for all of the ceramics. The thermal stability of the radiation defects was investigated by isochronal annealing up to 1050 °C.